DCL-16-090, Supplement to License Amendment Request 15-03, Application of Alternative Source Term.

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Supplement to License Amendment Request 15-03, Application of Alternative Source Term.
ML16259A117
Person / Time
Site: Diablo Canyon  Pacific Gas & Electric icon.png
Issue date: 09/15/2016
From: Gerfen P
Pacific Gas & Electric Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-16-090, LAR 15-03
Download: ML16259A117 (21)
v • d • e


Text

Pacific Gas and Electric Company*

Paula Gerfen Diablo Canyon Power Plant Station Director Mail code 104/5/502 P.O. Box 56 Avila Beach, CA 93424 805.545.4596 Internal: 691.4596 Fax: 805.545.4234 September 15, 2016 PG&E Letter DCL-16-090 U.S. Nuclear Regulatory Commission 10 CFR 50.90 ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Supplement to License Amendment Request 15-03, "Application of Alternative Source Term"

References:

1. PG&E Letter DCL-15-069, "License Amendment Request 15-03,

'Application of Alternative Source Term,"' dated June 17, 2015 (ADAMS Accession No. ML15176A539)

2. PG&E Letter DCL-15-1 05, "Supplement to License Amendment Request 15-03, 'Application of Alternative Source Term,"' dated August 31, 2015 (ADAMS Accession No. ML15243A363)
3. PG&E Letter DCL-15-152, "Response to NRC Request for Additional Information Regarding License Amendment
4. PG&E Letter DCL-16-061, "Supplement to License Amendment Request 15-03, 'Application of Alternative Source Term,"' dated June 9, 2016 (ADAMS Accession No. ML16169A264)

Dear Commissioners and Staff:

License Amendment Request (LAR) 15-03, "Application of Alternative Source Term,"

was submitted by Pacific Gas and Electric Company (PG&E) Letter DCL-15-069 (Reference 1) and supplemented by PG&E Letters DCL-15-1 05 (Reference 2) and DCL-16-061 (Reference 4).

This letter further supplements LAR 15-03. Specifically, this letter provides a revised No Significant Hazards Consideration incorporating two changes since the original

  • LAR 15-03 submitted in Reference 1. This letter also provides a markup to replace the previously submitted markup of Technical Specification (TS) Bases Section 83.4.13. Finally, this letter withdraws the proposed change toTS Section 5.5.9, A member of the STARS Alliance
  • Callaway
  • Diablo Canyon
  • Palo Verde*
  • Wolf Creek

Document Control Desk PG&E Letter DCL-16-090 September 15, 2016 Page2 "Steam Generator (SG) Tube Inspection Program," that was included in Reference 1.

The first of the two changes incorporated in the revised No Significant Hazards Consideration was included in Reference 4. The change removes a proposed license condition requiring that a 2-inch gaseous radwaste system line that connects to the plant vent be classified as PG&E Design Class I. In Reference 4, an evaluation was provided along with a markup of the No Significant Hazards Consideration.

The second change from the original LAR 15-03 returns the accident-induced leakage performance criterion for the steam generator tube inspection program to its original value. The revision was previously made to make the criterion more restrictive; however, further evaluation determined that this change was not required for the Alternative Source Term LAR and conflicted with Steam Generator Tube Inspection requirements. Thus, a revised markup TS Bases Section B3.4.13 is provided. The proposed change toTS Section 5.5.9 that was included in Reference 1 is withdrawn. To support this change, two newTS Bases markup pages are provided. The two pages replace the corresponding pages provided in both Reference 1 and Reference 3.

This information does not affect the results of the technical assessment previously transmitted for LAR 15-03.

PG&E makes no new or revised regulatory commitments (as defined by NEI 99-04) in this letter.

If you have any questions or require additional information, please contact Mr. Hossein Hamzehee at (805) 545-4720.

I have been delegated the authority of James M. Welsch, Vice President- Nuclear Generation, during his absence. I declare under penalty of perjury that the foregoing is true and correct.

Executed on September 15, 2016.

Sinf& c; Paula Gerfen Station Director T

A member of the STARS Alliance Callaway

  • Diablo Canyon
  • Palo Verde*
  • Wolf Creek

Document Control Desk PG&E Letter DCL-16-090 September 15, 2016 Page 3 E 1d?/4418/50705089 Enclosure cc: Diablo Distribution cc/enc: Kriss M. Kennedy, NRC Region IV Administrator Chris W. Newport, NRC Senior Resident Inspector Gonzalo L. Perez, Branch Chief, California Dept. of Public Health Balwant K. Singal, NRR Project Manager A member of the STARS Alliance Callaway

  • Diablo Canyon
  • Palo Verde*
  • Wolf Creek

Enclosure PG&E Letter DCL-16-090 Enclosure : No Significant Hazards Consideration Analysis - Final : No Significant Hazards Consideration Analysis - Markup : Technical Specification Bases- Markup (For Information Only)

Enclosure Attachment 1 PG&E Letter DCL-16-090 - No Significant Hazards Consideration Analysis - Final

Enclosure Attachment 1 PG&E Letter DCL-16-090 4.3 No Significant Hazards Consideration As provided by 10 CFR 50.67, Pacific Gas & Electric (PG&E) is implementing the use of an Alternative Source Term (AST) and the dose calculation methodology described in Regulatory Guide (RG) 1.183 to calculate the accident doses to the Control Room, Technical Support Center (TSC), and offsite receptors following postulated design basis events that result in the release of radioactive material from reactor fuel at Diablo Canyon Power Plant (DCPP) Units 1 and 2. The AST and associated methodology for full implementation of AST define the amount, isotopic composition, physical and chemical characteristics, and timing of radioactive material releases following postulated events. Transport of the material to the Control Room, TSC, and offsite areas is modeled, and the resulting Total Effective Dose Equivalent (TEDE) is determined. Regulatory acceptance criteria account for the sum of the deep-dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures).

In accordance with 10 CFR 50.67(b), licensees wishing to adopt an AST must apply for a license amendment in accordance with 10 CFR 50.90.

In support of the revised analyses applying AST, the following Technical Specification (TS) changes are being made: the definition for Dose Equivalent lodine-131 (DEl) is revised to be consistent with AST dose conversion factor usage, the limit for reactor coolant system Dose Equivalent Xenon-133 (DEX) activity is decreased to control the noble gas activity in the coolant to levels below the design basis values, the requirement for containment penetrations is revised to require the 48-inch containment purge supply and exhaust valves to be sealed closed during operation MODES 1, 2, 3, and 4 eliminating a potential dose contribution release path, and the testing requirement for the auxiliary building ventilation system charcoal filter is also revised to be more restrictive.

Other changes to the TSs involve the adoption of terminology on which AST is based.

AST methods have been utilized in the analysis of the limiting design basis accidents, as follows: loss of coolant accident (LOCA), fuel handing accident (FHA) in the containment and in the fuel handling building, locked rotor accident (LRA), control rod ejection accident (CREA), main steam line break (MSLB), and steam generator tube rupture (SGTR). AST methods have also been utilized in the analysis of the limiting Condition II event, the loss of load (LOL) accident.

Other changes incorporated in the revised analyses include revising atmospheric dispersion factors (x/0), reducing the minimum decay time before fuel movement, adding shielding to the Control Room for additional protection of Control Room personnel and adding a high efficiency particulate air (HEPA) filter for additional protection of TSC personnel. In addition, a portion of the 40-inch Containment Penetration Area Ventilation line is being reclassified from PG&E Design Class II to PG&E Design Class I. Because AST methodologies better represent the physical characteristics and timing of the radionuclide release following a postulated LOCA, containment spray is now relied upon during the recirculation of sump water for continued removal of iodine and particulate from Page 1 of 6

Enclosure Attachment 1 PG&E Letter DCL-16-090 the containment atmosphere for spray duration (injection plus recirculation) greater than 6.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />. In addition, setpoint changes are being made to the Control Room intake radiation monitors to incorporate the effect of all possible release points from a FHA.

PG&E has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

This license amendment does not physically impact any system, structure, or component (SSC) that is a potential initiator of an accident. Therefore, implementation of AST, the AST assumptions and inputs, the proposed TS changes, and new yJQ values have no impact on the probability for initiation of any design basis accident. Once the occurrence of an accident has been postulated, the new accident source term and x!Q values are inputs to analyses that evaluate the radiological consequences of the postulated events.

Reactor coolant specific activity, testing criteria of charcoal filters, and the accident induced primary-to-secondary system leakage performance criterion are not initiators for any accident previously evaluated. The proposed change to require the 48-inch containment purge valves to be sealed closed during operating MODES 1, 2, 3, and 4 is not an accident initiator for any accident previously evaluated. The change in the classification of a portion of the 40-inch Containment Penetration Area Ventilation line is also not an accident initiator for any accident previously evaluated. Thus, the proposed TS changes and AST implementation will not increase the probability of an accident.

The change to the decay time prior to fuel movement is not an accident initiator. Decay time is used to determine the source term for the dose consequence calculation following a potential FHA and has no effect on the probability of the accident. Likewise, the change to the Control Room radiation monitors setpoint cannot cause an accident and the operation of containment spray during the recirculation phase is used for mitigation of a LOCA, and thus not an accident initiator.

As a result, there are no proposed changes to the parameters or conditions that could contribute to the initiation of an accident previously evaluated in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR). As such, the AST cannot affect the probability of an accident previously evaluated.

Page 2 of6

Enclosure Attachment 1 PG&E Letter DCL-16-090 Regarding accident consequences, equipment and components affected by the proposed changes are mitigative in nature and relied upon once the accident has been postulated. The license amendment implements a new calculation methodology for determining accident consequences and does not adversely affect any plant component or system that is credited to mitigate fuel damage. Subsequently, no conditions have been created that could significantly increase the consequences of any accidents previously evaluated.

Requiring that the 48-inch containment purge supply and exhaust valves be sealed closed during operating MODES 1, 2, 3, and 4 eliminates a potential path for radiological release following events that result in radioactive material releases to the containment, thus reducing potential consequences of the event. The auxiliary building ventilation system allowable methyl iodide penetration limit is being changed, which results in more stringent testing requirements, and thus higher filter efficiencies for reducing potential releases.

Changes to the operation of the containment spray system to require operation during the recirculation mode are also mitigative in nature.

While the plant design basis has always included the ability to implement containment spray during recirculation, this license amendment now requires operation of containment spray in the recirculation mode for dose mitigation. DCPP is designed and licensed to operate using containment spray in the recirculation mode. As such, operation of containment spray in the recirculation mode has already been analyzed, evaluated, and is currently controlled by Emergency Operating Procedures. Usage of recirculation spray reduces the consequence of the postulated event.

Likewise, the additional shielding to the Control Room and the addition of a HEPA filter to the TSC ventilation system reduces the consequences of the postulated event to the Control Room and TSC personnel. Lowering the limit for DEX lowers potential releases. By reclassifying a portion of the 40-inch Containment Penetration Area Ventilation line to PG&E Design Class I, this line will be seismically qualified, thus assuring that post-LOCA release points are the same as those used for determining x!Q values.

The change to the decay time from 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to fuel movement is an input to the FHA. Although less decay will result in higher released activity, the results of the FHA dose consequence analysis remain within the dose acceptance criteria of the event. Also, the radiation levels to an operator from a raised fuel assembly may increase due to a lower decay time, however, any exposure will continue to be maintained under 10 CFR 20 limits by the plant Radiation Protection Program.

Plant-specific radiological analyses have been performed using the AST methodology, assumption and inputs, as well as new x!Q values. The results of the dose consequences analyses demonstrate that the Page 3 of6

Enclosure Attachment 1 PG&E Letter DCL-16-090 regulatory acceptance criteria are met for each analyzed event.

Implementing the AST involves no facility equipment, procedure, or process changes that could significantly affect the radioactive material actually released during an event. Subsequently, no conditions have been created that could significantly increase the consequences of any of the events being evaluated.

Based on the above discussion, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different accident from any accident previously evaluated?

Response: No.

This license amendment does not alter or place any SSC in a configuration outside its design or analysis limits and does not create any new accident scenarios.

The AST methodology is not an accident initiator, as it is a method used to estimate resulting postulated design basis accident doses. The proposed TS changes reflect the plant configuration that supports implementation of the new methodology and supports reduction in dose consequences.

DCPP is designed and licensed to operate using containment spray in the recirculation mode. This change will not affect any operational aspect of the system or any other system, thus no new modes of operation are introduced by the proposed change.

The function of the radiation monitors has not changed; only the setpoint has changed as a result of an assessment of all potential release pathways. The continued operation of containment spray and the radiation monitor setpoint change do not create any new failure modes, alter the nature of events postulated in the UFSAR, nor introduce any unique precursor mechanism.

Requiring the 48-inch containment purge valves to be sealed closed during operating MODES 1, 2, 3, and 4 does not introduce any new accident precursor. This change only eliminates a potential release path for radionuclides following a LOCA.

The proposed TS testing criteria for the auxiliary building ventilation system charcoal filters cannot create an accident, but results in requiring more efficient filtration of potentially released iodine. The proposed changes to the DEX activity limit, the TS terminology, and the decay time of the fuel before movement are also unrelated to accident initiators.

Page 4 of6

Enclosure Attachment 1 PG&E Letter DCL-16-090 The only physical changes to the plant being made in support of AST is the addition of Control Room shielding in an area previously modified, the addition of a HEPA filter at the intake of the TSC normal ventilation system, and the upgrade to the damper actuators, pressure switches, and damper solenoid valves to support reclassifying a portion of the Containment Penetration Area Ventilation line to PG&E Design Class I.

Both Control Room shielding and HEPA filtration are mitigative in nature and do not have any impact on plant operation or system response following an accident. The Control Room modification for adding the shielding will meet applicable loading limits, so the addition of the shielding cannot initiate a failure. Upgrading damper actuators, pressure switches, and damper solenoid valves involve replacing existing components with components that are PG&E Design Class I. Therefore, the addition of shielding, a HEPA filter, and upgrading components cannot create a new or different kind of accident.

Since the function of the SSCs has not changed for AST implementation, no new failure modes are created by this proposed change. The AST change itself does not have the capability to initiate accidents.

Therefore, the proposed change does not create the possibility of a new or different type of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

Implementing the AST is relevant only to calculated dose consequences of potential design basis accidents evaluated in Chapter 15 of the UFSAR.

The changes proposed in this license amendment involve the use of a new analysis methodology and related regulatory acceptance criteria.

New atmospheric dispersion factors, which are based on site specific meteorological data, were calculated in accordance with regulatory guidelines. The proposed TS, TS Bases, and UFSAR changes reflect the plant configuration that will support implementation of the new methodology and result in operation in accordance with regulatory guidelines that support the revisions to the radiological analyses of the limiting design basis accidents. Conservative methodologies, per the guidance of RG 1.183, have been used in performing the accident analyses. The radiological consequences of these accidents are all within the regulatory acceptance criteria associated with the use of AST methodology.

The change to the minimum decay time prior to fuel movement results in higher fission product releases after a FHA. However, the results of the Page 5 of6

Enclosure Attachment 1 PG&E Letter DCL-16-090 FHA dose consequence analysis remain within the dose acceptance criteria of the event.

The proposed changes continue to ensure that the dose consequences of design basis accidents at the exclusion area, low population zone boundaries, in the TSC, and in the Control Room are within the corresponding acceptance criteria presented in RG 1.183 and 10 CFR 50.67. The margin of safety for the radiological consequences of these accidents is provided by meeting the applicable regulatory limits, which are set at or below the 10 CFR 50.67 limits. An acceptable margin of safety is inherent in these limits.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above evaluation, PG&E concludes that the proposed change does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Page 6 of6

Enclosure Attachment 2 PG&E Letter DCL-16-090 - No Significant Hazards Consideration Analysis - Markup

Enclosure Attachment 2 PG&E Letter DCL-16-090 4.3 No Significant Hazards Consideration As provided by 10 CFR 50.67, Pacific Gas & Electric (PG&E) is implementing the use of an Alternative Source Term (AST) and the dose calculation methodology described in Regulatory Guide (RG) 1.183 to calculate the accident doses to the Control Room, Technical Support Center (TSC), and offsite receptors following postulated design basis events that result in the release of radioactive material from reactor fuel at Diablo Canyon Power Plant (DCPP) Units 1 and 2. The AST and associated methodology for full implementation of AST define the amount, isotopic composition, physical and chemical characteristics, and timing of radioactive material releases following postulated events. Transport of the material to the Control Room, TSC, and offsite areas is modeled, and the resulting Total Effective Dose Equivalent (TEDE) is determined. Regulatory acceptance criteria account for the sum of the deep-dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures).

In accordance with 10 CFR 50. 67(b), licensees wishing to adopt an AST must apply for a license amendment in accordance with 10 CFR 50.90.

In support of the revised analyses applying AST, the following Technical Specification (TS) changes are being made: the definition for Dose Equivalent lodine-131 (DEl) is revised to be consistent with AST dose conversion factor usage, the limit for reactor coolant system Dose Equivalent Xenon-133 (DEX) activity is decreased to control the noble gas activity in the coolant to levels below the design basis values, the requirement for containment penetrations is revised to require the 48-inch containment purge supply and exhaust valves to be sealed closed during operation MODES 1, 2, 3, and 4 eliminating a potential dose contribution release path, the accident induced leakage performance criterion for the steam generator tube inspection program is revised to be more restrictive, and the testing requirement for the auxiliary building ventilation system charcoal filter is also revised to be more restrictive. Other changes to the TSs involve the adoption of terminology on which AST is based.

AST methods have been utilized in the analysis of the limiting design basis accidents, as follows: loss of coolant accident (LOCA), fuel handing accident (FHA) in the containment and in the fuel handling building, locked rotor accident (LRA), control rod ejection accident (CREA), main steam line break (MSLB), and steam generator tube rupture (SGTR). AST methods have also been utilized in the analysis of the limiting Condition II event, the loss of load (LOL) accident.

Other changes incorporated in the revised analyses include revising atmospheric dispersion factors (r./0), reducing the minimum decay time before fuel movement, adding shielding to the Control Room for additional protection of Control Room personnel and adding a high efficiency particulate air (HEPA) filter for additional protection of TSC personnel. In addition, a portion of the 40-inch Containment Penetration Area Ventilation line and a portion of the 2 inch gaseous radv1aste system line '.Yhich connect to the Plant \lent are is being reclassified from PG&E Design Class II to PG&E Design Class I. Because AST methodologies better represent the physical characteristics and timing of the Page 1 of 6

Enclosure Attachment 2 PG&E Letter DCL-16-090 radionuclide release following a postulated LOCA, containment spray is now relied upon during the recirculation of sump water for continued removal of iodine and particulate from the containment atmosphere for spray duration (injection plus recirculation) greater than 6.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />. In addition, setpoint changes are being made to the Control Room intake radiation monitors to incorporate the effect of all possible release points from a FHA.

PG&E has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

This license amendment does not physically impact any system, structure, or component (SSC) that is a potential initiator of an accident. Therefore, implementation of AST, the AST assumptions and inputs, the proposed TS changes, and new x!Q values have no impact on the probability for initiation of any design basis accident. Once the occurrence of an accident has been postulated, the new accident source term and x!Q values are inputs to analyses that evaluate the radiological consequences of the postulated events.

Reactor coolant specific activity, testing criteria of charcoal filters, and the accident induced primary-to-secondary system leakage performance criterion are not initiators for any accident previously evaluated. The proposed change to require the 48-inch containment purge valves to be sealed closed during operating MODES 1, 2, 3, and 4 is not an accident initiator for any accident previously evaluated. The change in the classifications of a portion of the 40-inch Containment Penetration Area Ventilation line and a portion of the 2 inch gaseous rad'Naste system line is also not an accident initiator for any accident previously evaluated.

Thus, the proposed TS changes and AST implementation will not increase the probability of an accident.

The change to the decay time prior to fuel movement is not an accident initiator. Decay time is used to determine the source term for the dose consequence calculation following a potential FHA and has no effect on the probability of the accident. Likewise, the change to the Control Room radiation monitors setpoint cannot cause an accident and the operation of containment spray during the recirculation phase is used for mitigation of a LOCA, and thus not an accident initiator.

As a result, there are no proposed changes to the parameters or conditions that could contribute to the initiation of an accident previously Page 2 of6

Enclosure Attachment 2 PG&E Letter DCL-16-090 evaluated in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR). As such, the AST cannot affect the probability of an accident previously evaluated.

Regarding accident consequences, equipment and components affected by the proposed changes are mitigative in nature and relied upon once the accident has been postulated. The license amendment implements a new calculation methodology for determining accident consequences and does not adversely affect any plant component or system that is credited to mitigate fuel damage. Subsequently, no conditions have been created that could significantly increase the consequences of any accidents previously evaluated.

Requiring that the 48-inch containment purge supply and exhaust valves be sealed closed during operating MODES 1, 2, 3, and 4 eliminates a potential path for radiological release following events that result in radioactive material releases to the containment, thus reducing potential consequences of the event. The steam generator tube inspection testing criterion for accident induced leakage is being changed , resulting in IO'Ner leakage rates , and thus less potential releases due to primary to secondary leakage. The auxiliary building ventilation system allowable methyl iodide penetration limit is being changed, which results in more stringent testing requirements, and thus higher filter efficiencies for reducing potential releases.

Changes to the operation of the containment spray system to require operation during the recirculation mode are also mitigative in nature.

While the plant design basis has always included the ability to implement containment spray during recirculation, this license amendment now requires operation of containment spray in the recirculation mode for dose mitigation. DCPP is designed and licensed to operate using containment spray in the recirculation mode. As such, operation of containment spray in the recirculation mode has already been analyzed, evaluated, and is currently controlled by Emergency Operating Procedures. Usage of recirculation spray reduces the consequence of the postulated event.

Likewise, the additional shielding to the Control Room and the addition of a HEPA filter to the TSC ventilation system reduces the consequences of the postulated event to the Control Room and TSC personnel. Lowering the limit for DEX lowers potential releases. By reclassifying a portion of the 40-inch Containment Penetration Area Ventilation line and a portion of the 2 inch gaseous radvJaste system line to PG&E Design Class I, these this lines will be seismically qualified, thus assuring that post-LOCA release points are the same as those used for determining x!O values.

The change to the decay time from 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to fuel movement is an input to the FHA. Although less decay will result in higher released activity, the results of the FHA dose consequence analysis Page 3 of6

Enclosure Attachment 2 PG&E Letter DC L-16-090 remain within the dose acceptance criteria of the event. Also, the radiation levels to an operator from a raised fuel assembly may increase due to a lower decay time, however, any exposure will continue to be maintained under 10 CFR 20 limits by the plant Radiation Protection Program.

Plant-specific radiological analyses have been performed using the AST methodology, assumption and inputs, as well as new xiQ values. The results of the dose consequences analyses demonstrate that the regulatory acceptance criteria are met for each analyzed event.

Implementing the AST involves no facility equipment, procedure, or process changes that could significantly affect the radioactive material actually released during an event. Subsequently, no conditions have been created that could significantly increase the consequences of any of the events being evaluated.

Based on the above discussion, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different accident from any accident previously evaluated?

Response: No.

This license amendment does not alter or place any SSC in a configuration outside its design or analysis limits and does not create any new accident scenarios.

The AST methodology is not an accident initiator, as it is a method used to estimate resulting postulated design basis accident doses. The proposed TS changes reflect the plant configuration that supports implementation of the new methodology and supports reduction in dose consequences.

DCPP is designed and licensed to operate using containment spray in the recirculation mode. This change will not affect any operational aspect of the system or any other system, thus no new modes of operation are introduced by the proposed change.

The function of the radiation monitors has not changed; only the setpoint has changed as a result of an assessment of all potential release pathways. The continued operation of containment spray and the radiation monitor setpoint change do not create any new failure modes, alter the nature of events postulated in the UFSAR, nor introduce any unique precursor mechanism.

Requiring the 48-inch containment purge valves to be sealed closed during operating MODES 1, 2, 3, and 4 does not introduce any new Page 4 of 6

Enclosure Attachment 2 PG&E Letter DCL-16-090 accident precursor. This change only eliminates a potential release path for radionuclides following a LOCA.

The proposed TS testing criteria for the auxiliary building ventilation system charcoal filters and the proposed performance criteria for steam generator tube integrity also cannot create an accident, but results in requiring more efficient filtration of potentially released iodin ~HH-~s allo'Nable primary to secondary leakage . The proposed changes to the DEX activity limit, the TS terminology, and the decay time of the fuel before movement are also unrelated to accident initiators.

The only physical changes to the plant being made in support of AST is the addition of Control Room shielding in an area previously modified, the addition of a HEPA filter at the intake of the TSC normal ventilation system, and the upgrade to the damper actuators, pressure switches, and damper solenoid valves to support reclassifying a portion of the Containment Penetration Area Ventilation line to PG&E Design Class I.

Both Control Room shielding and HEPA filtration are mitigative in nature and do not have any impact on plant operation or system response following an accident. The Control Room modification for adding the shielding will meet applicable loading limits, so the addition of the shielding cannot initiate a failure. Upgrading damper actuators, pressure switches, and damper solenoid valves involve replacing existing components with components that are PG&E Design Class I. Therefore, the addition of shielding, a HEPA filter, and upgrading components cannot create a new or different kind of accident.

Since the function of the SSCs has not changed for AST implementation, no new failure modes are created by this proposed change. The AST change itself does not have the capability to initiate accidents.

Therefore, the proposed change does not create the possibility of a new or different type of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

Implementing the AST is relevant only to calculated dose consequences of potential design basis accidents evaluated in Chapter 15 of the UFSAR.

The changes proposed in this license amendment involve the use of a new analysis methodology and related regulatory acceptance criteria.

New atmospheric dispersion factors, which are based on site specific meteorological data, were calculated in accordance with regulatory guidelines. The proposed TS, TS Bases, and UFSAR changes reflect the plant configuration that will support implementation of the new Page 5 of6

Enclosure Attachment 2 PG&E Letter DCL-16-090 methodology and result in operation in accordance with regulatory guidelines that support the revisions to the radiological analyses of the limiting design basis acCidents. Conservative methodologies, per the guidance of RG 1.183, have been used in performing the accident analyses. The radiological consequences of these accidents are all within the regulatory acceptance criteria associated with the use of AST methodology.

The change to the minimum decay time prior to fuel movement results in higher fission product releases after a FHA. However, the results of the FHA dose consequence analysis remain within the dose acceptance criteria of the event.

The proposed changes continue to ensure that the dose consequences of design basis accidents at the exclusion area, low population zone boundaries, in the TSC, and in the Control Room are within the corresponding acceptance criteria presented in RG 1.183 and 10 CFR 50.67. The margin of safety for the radiological consequences of these accidents is provided by meeting the applicable regulatory limits, which are set at or below the 10 CFR 50.67 limits. An acceptable margin of safety is inherent in these limits.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above evaluation, PG&E concludes that the proposed change does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Page 6 of6

Enclosure Attachment 3 PG&E Letter DCL-16-090 - Technical Specification Bases - Markup (For Information Only)

RCS Operational LEAKAGE B 3.4.13 BASES (continued)

APPLICABLE Except for primary to secondary LEAKAGE, the safety analyses do not SAFETY address operational LEAKAGE. However, other operational LEAKAGE ANALYSES is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. Safety analyses for design basis events that model primary to secondary LEAKAGE result in steam discharge to the atmosphere. The safety analysis for the SLB event assumes that primary to secondary LEAKAGE is 10.5 gpm (room temperature conditions) from the faulted SG or increases to 10.5 gpm as a result of accident induced conditions, and 0.1 gpm (room temperature conditions) from each intact SG. The safety analyses for events resulting in steam discharge to the atmosphere, other than SGTR and SLB , assume that primary to secondary LEAKAGE from all SGs is 0.75 gpm (hot conditionsStandard Temperature and Pressure) under accident conditions. For conservativism. the SLB analysis assumes that the total 0. 75 gpm tube leakage is assigned to the faulted steam generator. The LCO requirement to limit primary to secondary LEAKAGE through any one SG to less than or equal to 150 gallons per day is significantly less than the conditions assumed in the SLB safety analysis for the faulted SG.

Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident.

To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). The leakage contaminates the secondary fluid.

The SGTR (Ref. 3) is more limiting for radiological releases at the site boundary. The radiological dose analysis assumes loss of off-site power at the time of reactor trip with no subsequent condenser cooling available. The steam generator (SG) PORV for the SG that has sustained the tube rupture is assumed to fail open for 30 minutes, at which time the operator closes the block valve to the PORV. The dose consequences resulting from the SGTR accident are within the limits defined in 10 CFR ~50.67 (Ref. 6).

The SLB is more limiting for site radiation releases for events other than SGTR. The safety analysis for the SLB accident assumes 10.5 gpm primary to secondary LEAKAGE is through the faulted SG.

The dose consequences resulting from the SLB accident ar within the limits defined in 10 CFR ~50.67or the staff approved licensing basis (i.e., small fraction of these limits) . '

The safety analysis for RCS main loop piping for GDC 4, 1987 (Ref. 1) assumes 1 gpm unidentified leakage and monitoring per RG 1.45 (Ref. 2) are maintained (Ref. 4 and 5).

The RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

(continued)

DIABLO CANYON- UNITS 1 & 2 Rev 90 Page 71 of 108 Tab_B3!4u3r09.doc 0913.1047

Steam Generator (SG) Tube Integrity B 3.4.17 BASES LCO Structural integrity requires that the primary membrane stress intensity (continued) in a tube not exceed the yield strength for all ASME Code, Section Ill, Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the design specification. This includes safety factors and applicable design basis loads based on ASME Code, Section Ill, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121 (Ref. 5).

The accident induced leakage performance criterion ensures (a) that the primary to secondary LEAKAGE caused by a design basis accident, other than a SGTR, is within the accident analysis assumptions, and (b) that the primary to secondary LEAKAGE will not exceed 1 gpm per SG (except for specific types of degradation at specific locations where the NRC has approved greater accident induced leakage) to ensure that the potential for induced leakage during severe accidents will be maintained at a level that will not increase risk. The accident analysis for the SLB e);ent assumes that accident induced leakage does not e~<eeed 10.5 gpm in the faulted SG and 0.1 gpm in each intact SG . For the faulted SG in the SLB event, 10.5 gpm is the accident induced leakage limit, of which no more than 1 gpm can come from sources not specifically exempted by the NRC from this 1 gpm limit. The accident analyses for events resulting in steam release to the atmosphere. other than SGTR..~. and SLB assume that leakage does not exceed 0.75 gpm total under accident conditions.

For conservativism. the SLB analysis assumes that the total 0.75 gpm tube leakage is assigned to the faulted steam generator. The accident induced leakage rate includes any primary to secondary LEAKAGE existing prior to the accident in addition to primary to secondary LEAKAGE induced during the accident.

The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation.

The limit on operational LEAKAGE is contained in LCO 3.4.13, "RCS Operational LEAKAGE," and limits primary to secondary LEAKAGE through any one SG to 150 gallons per day. This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.

APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1, 2, 3, or 4.

RCS conditions are far less challenging in MODES 5 and 6 than during MODES 1, 2, 3, and 4. In MODES 5 and 6, primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.

(continued)

DIABLO CANYON -UNITS 1 & 2 Rev 9D Page 105 of 108 Tab_B3!4u3r09.doc 0913.1047

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